Printed circuit heat exchangers (PCHEs) are a promising technology for the closed Brayton cycle of nuclear energy due to their high compactness, capability to withstand high pressure and large temperature difference, and high heat transfer efficiency. The design calculation of PCHEs has a significant impact on the cycle performance. This paper proposes a theoretical model for predicting the thermal-hydraulic performance of semicircular straight-channel PCHEs. The irregular fins are equivalent to rectangular fins of equal height, making the model more simplified and accurate. The model is verified using experimental data from the literature. Based on this model, the heat transfer characteristics of composite PCHEs, where crossflow exists at the inlet and outlet, are investigated. The results indicate that the heat transfer performance of PCHEs can be more accurately predicted by using the length of the overlapping region of hot and cold channels as the reference length. Furthermore, the effects of geometric parameters on the heat transfer performance of semicircular straight-channel PCHEs are studied. The results show that the heat transfer performance of the semicircular straight-channel PCHEs decreases with the increase of the channel radius and the vertical interval, and increases with the increase of the horizontal interval. And the horizontal and vertical interval affect heat transfer performance slightly. This study provides a reference for the design and application of PCHEs in the helium closed Brayton cycle in the future.
{"title":"Investigation of Design Models for Printed Circuit Heat Exchangers With Straight Semicircular Channels","authors":"Junjun Mao, Gang Zhao, Xiaoyong Yang","doi":"10.1115/icone29-93325","DOIUrl":"https://doi.org/10.1115/icone29-93325","url":null,"abstract":"\u0000 Printed circuit heat exchangers (PCHEs) are a promising technology for the closed Brayton cycle of nuclear energy due to their high compactness, capability to withstand high pressure and large temperature difference, and high heat transfer efficiency. The design calculation of PCHEs has a significant impact on the cycle performance. This paper proposes a theoretical model for predicting the thermal-hydraulic performance of semicircular straight-channel PCHEs. The irregular fins are equivalent to rectangular fins of equal height, making the model more simplified and accurate. The model is verified using experimental data from the literature. Based on this model, the heat transfer characteristics of composite PCHEs, where crossflow exists at the inlet and outlet, are investigated. The results indicate that the heat transfer performance of PCHEs can be more accurately predicted by using the length of the overlapping region of hot and cold channels as the reference length. Furthermore, the effects of geometric parameters on the heat transfer performance of semicircular straight-channel PCHEs are studied. The results show that the heat transfer performance of the semicircular straight-channel PCHEs decreases with the increase of the channel radius and the vertical interval, and increases with the increase of the horizontal interval. And the horizontal and vertical interval affect heat transfer performance slightly. This study provides a reference for the design and application of PCHEs in the helium closed Brayton cycle in the future.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"12 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123920951","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A Phased Mission System (PMS) is defined as a system subject to multiple, consecutive and non-overlapping operation phases of tasks during its mission, which is commonly found in complex technological and engineering practices such as aerospace, nuclear and military operations, etc. Over the past few decades, extensive efforts have been devoted to advance the theories, methods, techniques and tools used for reliability analysis of PMS. The general methods include fault tree based combinatorial approaches and Markov chains. In this paper, a success-oriented GO-FLOW method with a new exact algorithm is presented for reliability analysis of multi-phase mission systems. The feasibility and correctness of the extended GO-FLOW method (GFA) for PMS analysis are proved by comparing the results with fault tree analysis based on Sum of Disjoint Products (SDP) generation algorithm for two example case studies. The comparison results show that the GO-FLOW method can be effectively applied for reliability analysis of PMS in a very compact way. Consistent results can be obtained using extended GO-FLOW algorithm when compared to the fault tree analysis.
{"title":"Reliability Analysis of Phased Mission Systems Using GO-FLOW Methodology","authors":"Zhan-Xiang He, Jun Yang, T. Matsuoka, Ming Yang","doi":"10.1115/icone29-91772","DOIUrl":"https://doi.org/10.1115/icone29-91772","url":null,"abstract":"\u0000 A Phased Mission System (PMS) is defined as a system subject to multiple, consecutive and non-overlapping operation phases of tasks during its mission, which is commonly found in complex technological and engineering practices such as aerospace, nuclear and military operations, etc. Over the past few decades, extensive efforts have been devoted to advance the theories, methods, techniques and tools used for reliability analysis of PMS. The general methods include fault tree based combinatorial approaches and Markov chains. In this paper, a success-oriented GO-FLOW method with a new exact algorithm is presented for reliability analysis of multi-phase mission systems. The feasibility and correctness of the extended GO-FLOW method (GFA) for PMS analysis are proved by comparing the results with fault tree analysis based on Sum of Disjoint Products (SDP) generation algorithm for two example case studies. The comparison results show that the GO-FLOW method can be effectively applied for reliability analysis of PMS in a very compact way. Consistent results can be obtained using extended GO-FLOW algorithm when compared to the fault tree analysis.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128630975","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
With the development of artificial intelligence technology, various achievements have been realized in data-driven nuclear power plant fault diagnosis. Even endowed with high flexibility and practicability, most of the proposed data-driven methods are based on the same assumptions that the test data is in the same distribution as the training data. In practice, nuclear power plants may be in variable operating conditions, which brings challenges to the generalization of the diagnosis model trained by finite data. In this paper, the widely used data-driven models in nuclear power plant fault diagnosis: Random Forest (RF), K-Nearest Neighbor algorithm (KNN), Fully Connected Neural Network (FCNN) and Convolutional Neural Network (CNN) are taken as examples to study the influence of the distribution discrepancy between training data (source domain) and test data (target domain) on their generalization. The results show that the distribution discrepancy exert serious adverse effects on the diagnostic performance of the data-driven models. At the same time, to improve the generalization of data-driven models, a nuclear power plant fault diagnosis transfer learning method based on pre-trained model is proposed, which can utilize the fault diagnosis knowledge from the source domain task to accelerate the model training in the target domain task, so that the model can achieve better diagnosis performance with limited labeled data in target domain.
{"title":"Research on Generalization of Typical Data-Driven Fault Diagnosis Methods for Nuclear Power Plants","authors":"Jiangkuan Li, Meng Lin","doi":"10.1115/icone29-88934","DOIUrl":"https://doi.org/10.1115/icone29-88934","url":null,"abstract":"\u0000 With the development of artificial intelligence technology, various achievements have been realized in data-driven nuclear power plant fault diagnosis. Even endowed with high flexibility and practicability, most of the proposed data-driven methods are based on the same assumptions that the test data is in the same distribution as the training data. In practice, nuclear power plants may be in variable operating conditions, which brings challenges to the generalization of the diagnosis model trained by finite data. In this paper, the widely used data-driven models in nuclear power plant fault diagnosis: Random Forest (RF), K-Nearest Neighbor algorithm (KNN), Fully Connected Neural Network (FCNN) and Convolutional Neural Network (CNN) are taken as examples to study the influence of the distribution discrepancy between training data (source domain) and test data (target domain) on their generalization. The results show that the distribution discrepancy exert serious adverse effects on the diagnostic performance of the data-driven models. At the same time, to improve the generalization of data-driven models, a nuclear power plant fault diagnosis transfer learning method based on pre-trained model is proposed, which can utilize the fault diagnosis knowledge from the source domain task to accelerate the model training in the target domain task, so that the model can achieve better diagnosis performance with limited labeled data in target domain.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114287773","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Due to the advantages of compactness and enhanced heat transfer efficiency, the Helical-Coiled Once-Through Steam Generator (HCOTSG) is one of the best choices to be applied in the Lead-bismuth Fast Reactor (LFR). In order to investigate the flow and heat transfer characteristics in the shell side of lead-bismuth eutectic (LBE) HCOTSG for the purpose of safety analysis and optimization design, a CFD method for calculating the heat transfer and flow resistance in helical-coiled tube bundles is proposed. Based on available relevant experimental studies, the numerical model is validated. Validation results indicate that the liquid LBE cross flow through helical-coiled tube bundles can be well simulated using SST k-ω model and turbulent Prandtl number model. The maximum deviation for both friction factor and Nusselt number is less than 15%. After validating the reliability of the numerical model, geometry models with different helix angles of the shell side are established. Effects of helix angle on the flow resistance and heat transfer performances are presented. The flow and heat transfer characteristics in the helical-coiled tube bundle are numerically simulated and analyzed. This work provides a new method for studying the flow and heat transfer characteristics and optimization design of liquid metal H-OTSG.
{"title":"Numerical Investigations of the LBE Flow and Heat Transfer Characteristics in the Helical-Coiled Tube Bundle","authors":"Cong Shen, Limin Liu, Maolong Liu, H. Gu","doi":"10.1115/icone29-91911","DOIUrl":"https://doi.org/10.1115/icone29-91911","url":null,"abstract":"\u0000 Due to the advantages of compactness and enhanced heat transfer efficiency, the Helical-Coiled Once-Through Steam Generator (HCOTSG) is one of the best choices to be applied in the Lead-bismuth Fast Reactor (LFR). In order to investigate the flow and heat transfer characteristics in the shell side of lead-bismuth eutectic (LBE) HCOTSG for the purpose of safety analysis and optimization design, a CFD method for calculating the heat transfer and flow resistance in helical-coiled tube bundles is proposed. Based on available relevant experimental studies, the numerical model is validated. Validation results indicate that the liquid LBE cross flow through helical-coiled tube bundles can be well simulated using SST k-ω model and turbulent Prandtl number model. The maximum deviation for both friction factor and Nusselt number is less than 15%. After validating the reliability of the numerical model, geometry models with different helix angles of the shell side are established. Effects of helix angle on the flow resistance and heat transfer performances are presented. The flow and heat transfer characteristics in the helical-coiled tube bundle are numerically simulated and analyzed. This work provides a new method for studying the flow and heat transfer characteristics and optimization design of liquid metal H-OTSG.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126075070","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Wenjie Chen, Xianan Du, Youqi Zheng, Yongping Wang, Rong Wang, Hongchun Wu
NECP-SARAX is a neutronics code system for fast spectrum reactor developed by Nuclear Engineering Computational Physics Laboratory team of Xi'an Jiaotong University. In previous work, NECP-SARAX has shown high performance on fast spectrum reactor analysis. Recently, neutron-moderating materials are employed in advance reactors design where the pure fast spectrum is softened to intermediate and thermal energy spectrum. Due to the larger fission cross-section below the fast energy range, the volume of reactors reaching criticality can be reduced. Compared that in fast energy range, the temperature reactivity negative feedback resulting from the Doppler effect in thermal spectrum range is more significant, which is conducive to the safety and miniaturization of the reactors. To meet the design requirement of this kind of reactor, the assembly-wise neutron spectrum calculation module TULIP of NECP-SARAX has recently been extended to generate the cross sections for both the thermal and fast spectrum reactor system. Therefore, in this paper, the validation works of TULIP code have been performed. In order to systematically validate the accuracy of TULIP code, a series of benchmarks with neutron-moderating material are selected from the ICSBEP, such as HEU-MET-FAST-001-002,HEU-MET-FAST-027-001, U233-SOL-THERM-015-001. The numerical results showed that the TULIP code had accurate neutron spectrum calculation capability for the advanced nuclear reactor design.
{"title":"Validation of TULIP via ICSBEP Critical Benchmark","authors":"Wenjie Chen, Xianan Du, Youqi Zheng, Yongping Wang, Rong Wang, Hongchun Wu","doi":"10.1115/icone29-93353","DOIUrl":"https://doi.org/10.1115/icone29-93353","url":null,"abstract":"\u0000 NECP-SARAX is a neutronics code system for fast spectrum reactor developed by Nuclear Engineering Computational Physics Laboratory team of Xi'an Jiaotong University. In previous work, NECP-SARAX has shown high performance on fast spectrum reactor analysis. Recently, neutron-moderating materials are employed in advance reactors design where the pure fast spectrum is softened to intermediate and thermal energy spectrum. Due to the larger fission cross-section below the fast energy range, the volume of reactors reaching criticality can be reduced. Compared that in fast energy range, the temperature reactivity negative feedback resulting from the Doppler effect in thermal spectrum range is more significant, which is conducive to the safety and miniaturization of the reactors. To meet the design requirement of this kind of reactor, the assembly-wise neutron spectrum calculation module TULIP of NECP-SARAX has recently been extended to generate the cross sections for both the thermal and fast spectrum reactor system. Therefore, in this paper, the validation works of TULIP code have been performed. In order to systematically validate the accuracy of TULIP code, a series of benchmarks with neutron-moderating material are selected from the ICSBEP, such as HEU-MET-FAST-001-002,HEU-MET-FAST-027-001, U233-SOL-THERM-015-001. The numerical results showed that the TULIP code had accurate neutron spectrum calculation capability for the advanced nuclear reactor design.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"39 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126518226","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
There are many multiphase flow phenomena in steam generators (SG) of nuclear power plants (NPPs) and the movement of droplets affects the separation efficiency of dryers in SG. And the most important factor affecting droplet movement is drag force on droplet. Drag force coefficient (DRC) and deformation coefficient (DEC) as characterization coefficient of drag force have important research significance. In this paper, water and silicone oil are used as continuous phase and dispersed phase, respectively. Study on variation of the DRC when Re is in the range of 30–1200 is carried out. Firstly, the motion of droplets (MDs) of different sizes is visually studied by using the high-speed camera and images of MD is acquired. The droplet contour is recognized based on principle of boundary differentiation and the centroid coordinates of the droplet are determined, thus DRC and DEC are obtained. Besides, the relationship between DRC and DEC are researched. Additionally, Lattice Boltzmann Method (LBM) is used to simulate the droplet motion. The simulation results are compared with the experimental results, thus verifying the feasibility of LBM method to simulate the MD. The results show that when velocity of droplet (VD) is low, DRC is inversely proportional to VD. While VD is high, DRC is constant. When the droplet diameter (DD) is small, the final VD is proportional to DD, and when DD is large, the final VD is proportional to the square of DD. The DEC is linearly related to We. The larger the We, the smaller the DEC. Shan-Chen model of LBM is feasible for droplets with low Reynolds number (Re), while the simulation of droplets with high Re is the future prospect.
{"title":"Experimental and Numerical Research on Droplet Drag Coefficient and Deformation Coefficient in Nuclear Power Plants","authors":"Ru Li, Ruifeng Tian, Bowen Chen, Bo Wang","doi":"10.1115/icone29-90656","DOIUrl":"https://doi.org/10.1115/icone29-90656","url":null,"abstract":"\u0000 There are many multiphase flow phenomena in steam generators (SG) of nuclear power plants (NPPs) and the movement of droplets affects the separation efficiency of dryers in SG. And the most important factor affecting droplet movement is drag force on droplet. Drag force coefficient (DRC) and deformation coefficient (DEC) as characterization coefficient of drag force have important research significance. In this paper, water and silicone oil are used as continuous phase and dispersed phase, respectively. Study on variation of the DRC when Re is in the range of 30–1200 is carried out. Firstly, the motion of droplets (MDs) of different sizes is visually studied by using the high-speed camera and images of MD is acquired. The droplet contour is recognized based on principle of boundary differentiation and the centroid coordinates of the droplet are determined, thus DRC and DEC are obtained. Besides, the relationship between DRC and DEC are researched. Additionally, Lattice Boltzmann Method (LBM) is used to simulate the droplet motion. The simulation results are compared with the experimental results, thus verifying the feasibility of LBM method to simulate the MD. The results show that when velocity of droplet (VD) is low, DRC is inversely proportional to VD. While VD is high, DRC is constant. When the droplet diameter (DD) is small, the final VD is proportional to DD, and when DD is large, the final VD is proportional to the square of DD. The DEC is linearly related to We. The larger the We, the smaller the DEC. Shan-Chen model of LBM is feasible for droplets with low Reynolds number (Re), while the simulation of droplets with high Re is the future prospect.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125142786","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In the upgrades and innovations of nuclear fuel material, the U-50 wt.% Zr alloy is regarded as one of the most promising metallic fuel materials, due to its excellent thermal response, acceptable irradiation performance, and ease of fuel recycling. Under in-pile irradiation, large temperature gradients and dimensional changes contribute to complicated fuel thermal-mechanical behaviors, including the pore effect induced by fission gas production. However, the deficiency of the physical parameters of the porous U-50 wt.% Zr alloy makes it hardly possible to conduct fuel performance prediction under high irradiation conditions. To obtain Young’s modulus and thermal conductivity of porous U-50 wt.% Zr alloy, the molecular dynamics (MD) code LAMMPS and the modified embedded atom method (MEAM) potential for binary U-Zr system were incorporated. In this study, three-dimensional elastic constants were calculated by engineering strain loading method at different ambient temperatures and porosities, and the effective Young’s modulus was computed via Voigt averaging scheme. The phonon thermal conductivity was simulated with the Non-Equilibrium Molecular Dynamics (NEMD) method, and the electron thermal conductivity was predicted by semi-empirical correlations and existing density functional theory (DFT) results. The parallel model, series model and effective medium theory (EMT) were adopted to consider the mixture and pores effect. Finally, porosity factors were proposed to establish new semi-empirical correlations, which could give a preliminary prediction of Young’s modulus and thermal conductivity for porous U-50 wt.% Zr alloy.
{"title":"Preliminary Prediction of Young’s Modulus and Thermal Conductivity of Porous U-50 wt.% Zr Alloy: A Molecular Dynamic and Semi-Empirical Study","authors":"Mengke Cai, Tenglong Cong, H. Gu","doi":"10.1115/icone29-91602","DOIUrl":"https://doi.org/10.1115/icone29-91602","url":null,"abstract":"\u0000 In the upgrades and innovations of nuclear fuel material, the U-50 wt.% Zr alloy is regarded as one of the most promising metallic fuel materials, due to its excellent thermal response, acceptable irradiation performance, and ease of fuel recycling. Under in-pile irradiation, large temperature gradients and dimensional changes contribute to complicated fuel thermal-mechanical behaviors, including the pore effect induced by fission gas production. However, the deficiency of the physical parameters of the porous U-50 wt.% Zr alloy makes it hardly possible to conduct fuel performance prediction under high irradiation conditions. To obtain Young’s modulus and thermal conductivity of porous U-50 wt.% Zr alloy, the molecular dynamics (MD) code LAMMPS and the modified embedded atom method (MEAM) potential for binary U-Zr system were incorporated. In this study, three-dimensional elastic constants were calculated by engineering strain loading method at different ambient temperatures and porosities, and the effective Young’s modulus was computed via Voigt averaging scheme. The phonon thermal conductivity was simulated with the Non-Equilibrium Molecular Dynamics (NEMD) method, and the electron thermal conductivity was predicted by semi-empirical correlations and existing density functional theory (DFT) results. The parallel model, series model and effective medium theory (EMT) were adopted to consider the mixture and pores effect. Finally, porosity factors were proposed to establish new semi-empirical correlations, which could give a preliminary prediction of Young’s modulus and thermal conductivity for porous U-50 wt.% Zr alloy.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"19 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128861841","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yanyi Shen, Hongli Chen, Tao Ding, Tianyi Liu, Junjie Tang
It is of necessity and importance for the development of a real-time visualization and simulation platform for the lead-cooled fast reactor in order to provide a better learning and research platform for technicians. In this research, a visualization platform for the primary loop system of medium-sized modular lead-cooled fast reactor M2LFR-1000 has been developed based on Generic Platform Interface ICoCo and Qt framework, relying on the system code Relap5. The simulation code used in the visualization platform is first wrapped by the generic platform interface ICoCo (Interface for Code Coupling) and then compiled to a shared library. A multithreading C++ script is developed as the supervisor, which supervises the system code Relap5 and realizes the control of real-time simulation. The graphical man-machine interface of the platform is developed by a set of UI elements provided by Qt Widgets Module. The communication between simulation code and GUI is performed with the signals and slots mechanism, which requires a Qt’s C++ extension developed as the meta-object compiler (moc). Via the actual operation by the visualization and simulation platform, the results verify that the platform can realize the real-time simulation and control of the primary loop system of M2LFR-1000 and provide a practical means of real-time monitoring and regulation of reactor operations for technicians.
开发一个铅冷快堆的实时可视化仿真平台，为技术人员提供一个更好的学习和研究平台，是十分必要和重要的。本研究基于通用平台接口ICoCo和Qt框架，依托系统代码Relap5，开发了中型模块化铅冷快堆M2LFR-1000主回路系统可视化平台。可视化平台中使用的仿真代码首先由通用平台接口ICoCo (interface for code Coupling)封装，然后编译成共享库。开发多线程c++脚本作为监控程序，对系统代码Relap5进行监控，实现实时仿真的控制。平台的图形人机界面是由Qt Widgets Module提供的一组UI元素开发的。仿真代码和GUI之间的通信是通过信号和插槽机制来实现的，这需要Qt的c++扩展作为元对象编译器(moc)来开发。通过可视化仿真平台的实际运行，验证了该平台能够实现M2LFR-1000一次回路系统的实时仿真与控制，为技术人员实时监控和调控反应堆运行提供了一种实用的手段。
{"title":"Development and Application of a Real-Time Visualization and Simulation Platform Based on the Generic Platform Interface ICoCo and the Qt Framework","authors":"Yanyi Shen, Hongli Chen, Tao Ding, Tianyi Liu, Junjie Tang","doi":"10.1115/icone29-91644","DOIUrl":"https://doi.org/10.1115/icone29-91644","url":null,"abstract":"\u0000 It is of necessity and importance for the development of a real-time visualization and simulation platform for the lead-cooled fast reactor in order to provide a better learning and research platform for technicians. In this research, a visualization platform for the primary loop system of medium-sized modular lead-cooled fast reactor M2LFR-1000 has been developed based on Generic Platform Interface ICoCo and Qt framework, relying on the system code Relap5. The simulation code used in the visualization platform is first wrapped by the generic platform interface ICoCo (Interface for Code Coupling) and then compiled to a shared library. A multithreading C++ script is developed as the supervisor, which supervises the system code Relap5 and realizes the control of real-time simulation. The graphical man-machine interface of the platform is developed by a set of UI elements provided by Qt Widgets Module. The communication between simulation code and GUI is performed with the signals and slots mechanism, which requires a Qt’s C++ extension developed as the meta-object compiler (moc). Via the actual operation by the visualization and simulation platform, the results verify that the platform can realize the real-time simulation and control of the primary loop system of M2LFR-1000 and provide a practical means of real-time monitoring and regulation of reactor operations for technicians.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"64 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132457387","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
As one of the cooling methods for the blanket in a fusion reactor, forced convection using helium gas has been considered as one of the promising candidates. In this study, in order to clarify the effect of tube length on heat transfer for forced flow of helium gas under various velocities and pressures, experiments on forced convection heat transfer for helium gas in a small diameter tube were conducted. A circular tube made of platinum with an inner diameter of 1.8 mm and a heated length of 50 mm was used in this experiment. The tube was heated with exponentially increasing heat inputs. As a result of the experiment, the heat transfer process can be considered in the quasi-steady state when the e-folding time is larger than about 1.5 s. In addition, the heat transfer coefficient increased with the increases in velocity and pressure. The heat transfer was also higher than that of conventional turbulent heat transfer correlation. By comparing with the experimental results of tube with different heated lengths, it was found that the heat transfer coefficients for the heated length of 50 mm were higher than those of the one with a length of 90 mm.
{"title":"Heat Transfer for Forced Flow of Helium Gas in a Small Diameter Tube With Different Heated Length","authors":"Yushi Honjo, Feng Xu, Qiusheng Liu, M. Shibahara","doi":"10.1115/icone29-91899","DOIUrl":"https://doi.org/10.1115/icone29-91899","url":null,"abstract":"\u0000 As one of the cooling methods for the blanket in a fusion reactor, forced convection using helium gas has been considered as one of the promising candidates. In this study, in order to clarify the effect of tube length on heat transfer for forced flow of helium gas under various velocities and pressures, experiments on forced convection heat transfer for helium gas in a small diameter tube were conducted. A circular tube made of platinum with an inner diameter of 1.8 mm and a heated length of 50 mm was used in this experiment. The tube was heated with exponentially increasing heat inputs. As a result of the experiment, the heat transfer process can be considered in the quasi-steady state when the e-folding time is larger than about 1.5 s. In addition, the heat transfer coefficient increased with the increases in velocity and pressure. The heat transfer was also higher than that of conventional turbulent heat transfer correlation. By comparing with the experimental results of tube with different heated lengths, it was found that the heat transfer coefficients for the heated length of 50 mm were higher than those of the one with a length of 90 mm.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131075470","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The RISMC analysis method is a coupled dynamic safety analysis method that combines deterministic safety analysis and probabilistic safety analysis. It comprehensively simulates the dynamic response process of nuclear power plants under accidents through Monte Carlo sampling and dynamic event tree methods. In order to fully reflect the response process of the nuclear power plant and ensure the accuracy of the calculation results, it is necessary to sample a large number of input space parameters, which will result in unbearable time costs. Adaptive sampling can predict the state of the position input space parameters through a small number of forward sampling calculation results and divide the input space by generating limit surfaces to calculate the failure probability, which can greatly save computing time. Based on researching RISMC analysis tools at home and abroad, the research and development of adaptive sampling and RELAP5 coupling calculation software is realized. The test case of the Qinshan nuclear power plant station blackout accident shows that the adaptive sampling function of calculating failure space probability through a small amount of sampling is realized, and the efficiency of RISMC analysis and calculation is improved.
{"title":"Development of Adaptive Sampling Software Based on Relap5","authors":"Haoyin Chen, He Wang, Mohamedelmogtabh Omer Elfadni Suliman, Xinyue Wang, Qiang Zhao","doi":"10.1115/icone29-92454","DOIUrl":"https://doi.org/10.1115/icone29-92454","url":null,"abstract":"\u0000 The RISMC analysis method is a coupled dynamic safety analysis method that combines deterministic safety analysis and probabilistic safety analysis. It comprehensively simulates the dynamic response process of nuclear power plants under accidents through Monte Carlo sampling and dynamic event tree methods. In order to fully reflect the response process of the nuclear power plant and ensure the accuracy of the calculation results, it is necessary to sample a large number of input space parameters, which will result in unbearable time costs. Adaptive sampling can predict the state of the position input space parameters through a small number of forward sampling calculation results and divide the input space by generating limit surfaces to calculate the failure probability, which can greatly save computing time. Based on researching RISMC analysis tools at home and abroad, the research and development of adaptive sampling and RELAP5 coupling calculation software is realized. The test case of the Qinshan nuclear power plant station blackout accident shows that the adaptive sampling function of calculating failure space probability through a small amount of sampling is realized, and the efficiency of RISMC analysis and calculation is improved.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"36 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115147364","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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